• Nuclear Engineering and Technology
• /
• v.39 no.1
• /
• pp.63-74
• /
• 2007

In this paper, a reconfigurable controller consisting of a normal controller and a standby controller is designed to control the thermoelectric (TE) power in the SP-100 space reactor. The normal controller uses a model predictive control (MPC) method where the future TE power is predicted by using support vector regression. A genetic algorithm that can effectively accomplish multiple objectives is used to optimize the normal controller. The performance of the normal controller depends on the capability of predicting the future TE power. Therefore, if the prediction performance is degraded, the proportional-integral (PI) controller of the standby controller begins to work instead of the normal controller. Performance deterioration is detected by a sequential probability ratio test (SPRT). A lumped parameter simulation model of the SP-100 nuclear space reactor is used to verify the proposed reconfigurable controller. The results of numerical simulations to assess the performance of the proposed controller show that the TE generator power level controlled by the proposed reconfigurable controller could track the target power level effectively, satisfying all control constraints. Furthermore, the normal controller is automatically switched to the standby controller when the performance of the normal controller degrades.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.552-559
• /
• 2020

This paper presents an observer-based chattering free robust optimal control scheme to regulate the total power of a nuclear reactor. The non-linear model of nuclear reactor is linearized around a steady state operating point to obtain a linear model for which an optimal second order integral sliding mode controller is designed. A second order integral sliding mode observer is also designed to estimate the unmeasurable states. In order to avoid the chattering effect, the discontinuous input of both observer and controller are designed using the super-twisting algorithm. The proposed controller is realized by combining an optimal linear tracking controller with a second order integral sliding mode controller to ensure minimum control effort and robustness of the closed-loop system in the presence of uncertainties. The condition for the selection of gains of discontinuous control based on the super-twisting algorithm is derived using a strict Lyapunov function. Performance of the proposed observer based control scheme is demonstrated through non-linear simulation studies.

• Nuclear Engineering and Technology
• /
• v.52 no.7
• /
• pp.1443-1451
• /
• 2020

The objective of this study is to design a robust power control system for a small pressurized water reactor (PWR) to achieve stable power operations under conditions of external disturbances and internal model uncertainties. For this purpose, the multiple-input multiple-output transfer function models of the reactor core at five power levels are derived from point reactor kinetics equations and the Mann's thermodynamic model. Using the transfer function models, five local reactor power controllers are designed using an H infinity (H_∞) mixed sensitivity method to minimize the core power disturbance under various uncertainties at the five power levels, respectively. Then a multimodel approach with triangular membership functions is employed to integrate the five local controllers into a multimodel robust control system that is applicable for the entire power range. The performance of the robust power system is assessed against 10% of full power (FP) step load increase transients with coolant inlet temperature disturbances at different power levels and large-scope, rapid ramp load change transient. The simulation results show that the robust control system could maintain satisfactory control performance and good robustness of the reactor under external disturbances and internal model uncertainties, demonstrating the effective of the robust power control design.

• 제어로봇시스템학회:학술대회논문집
• /
• 1993.10a
• /
• pp.749-753
• /
• 1993

It is very difficult to control batch reactor with conventional linear controller due to its severe nonlinearity. To control the nonlinearity of batch reactor, we applied with relay feedback method and SOAS. The SOAS can be designed to work quite well, but it requires engineering effect and some knowledge about the process in order to get a satisfactory performance of the closed loop system For the applications to more reliable, further studies on robustness in various situations and process noises and would be required.

• Journal of the Korean Institute of Telematics and Electronics
• /
• v.11 no.3
• /
• pp.27-31
• /
• 1974

The maximum principle of Pontryagin provides the celebrated method to obtain the optimum control switching time and switching points on the nuclear reactor. The control trajectories transfered from its initial state to the target state are optimized based on time optioptimal control method with the given reactor parameters and the piecewise constant input values.

• 제어로봇시스템학회:학술대회논문집
• /
• 1999.10a
• /
• pp.252-255
• /
• 1999

Feed rate in the fed-batch reactor is the most important control variable in optimizing the reactor performance. Exact solution can be obtained only for limited cases of simple reactor. The complexity of the model equations makes it extremely difficult to solve fur the general class of system models. Evolutionary programming method is proposed to get the information of the profile types, and the final profile is calculated by that information. The modified evolutionary programming method is used to get the more optimal profiles and it is demonstrated that proposed method can solve a wide range of optimal control problems.

• 제어로봇시스템학회:학술대회논문집
• /
• 1999.10a
• /
• pp.45-48
• /
• 1999

Model predictive control algorithm requires a relevant model of the system to be controlled. Unfortunately, the first principle model describing a polymerization reaction system has a large number of parameters to be estimated. Thus there is a need for the identification and control of a polymerization reactor system by using available input-output data. In this work, the polynomial auto-regressive moving average (ARMA) models are employed as the input-output model and combined into the nonlinear model predictive control algorithm based on the successive linearization method. Simulations are conducted to identify the continuous styrene polymerization reactor system. The input variables are the jacket inlet temperature and the feed flow rate whereas the output variables are the monomer conversion and the weight-average molecular weight. The polynomial ARMA models obtained by the system identification are used to control the monomer conversion and the weight-average molecular weight in a continuous styrene polymerization reactor It is demonstrated that the nonlinear model predictive controller based on the polynomial ARMA model tracks the step changes in the setpoint satisfactorily. In conclusion, the polynomial ARMA model is proven effective in controlling the continuous styrene polymerization reactor.

• Proceedings of the ESK Conference
• /
• 1996.10a
• /
• pp.47-55
• /
• 1996

This study developed a comprehensive and easily applicable nuclear reactor control system evaluation method using reactor operators behavioral and mental workload database. A proposed control panel design cycle consists of the 5 steps: (1) finding out inconvenient, erroneous, and mentally stressful factors for the proposed design through evaluative experiments, (2) drafting improved design alternatives considering detective factors found out in the step (1), (3) comparative experiements for the design alternatives, (4) selecting a best design alternative, (5) returning to the step (1) and repeating the design cycle. Reactor operators behavioral and mental workload database collected from evaluative experiments in the step (1) and comparative experiments in the step (3) of the design cycle have a key roll in finding out defective factors and yielding the criteria for selection of the proposed reactor control systems. The behavioral database was designed to include the major informations about reactor operators' control behaviors: beginning time of operations, involved displays, classification of observational behaviors, dehaviors, decisions, involved control devices, classification of control behaviors, communications, emotional status, opinions for man-machine interface, and system event log. The database for mental workload scored from various physiological variables-EEG, EOG, ECG, and respir- ation pattern-was developed to indicate the most stressful situation during reactor control operations and to give hints for defective design factors. An experimental test for the evaluation method applied to the Compact Nuclear Simulator (CNS) installed in Korea Atomic Energy Research Institute (KAERI) suggested that some defective design factors of analog indicators should be improved and that automatization of power control to a target level would give relaxation to the subject operators in stressful situation.

• Proceedings of the KIEE Conference
• /
• 2007.07a
• /
• pp.1626-1627
• /
• 2007

This paper deals with the design and manufacturing of power cabinet for reactor power control system(PCS). The PCS provides the control signals and motive power to operate the CEDMs(Control Element Drive Mechanism). The CEDM is raise and lower the CEAs(Control Element Assemblies) in the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.4166-4178
• /
• 2021

The Control Rod Drive Mechanism (CRDM) is an essential part of the reactor, which realizes the start-stop and power adjustment of the reactor by lifting and lowering the control rod assembly. As a moving part in CRDM, the latch directly contacts with the control rod assembly, and the life of latch is closely related to the service life of the reactor. In this paper, the relationship between the life of the latch and the step stress, friction stress, and impact stress in the process of movement is analyzed, and the optimization methodology and process of latch life based on the approximate model are proposed. The design variables that affect the life of the latch are studied through the experimental design, and the optimization objective of design variables based on the latch life is established. Based on this, an approximate model of the life of the latch is built, and the multi-objective optimization of the life of the latch is optimized through the NSGA-II algorithm.